• Nuclear Engineering and Technology
• /
• v.18 no.3
• /
• pp.200-208
• /
• 1986

The loss of coolant accident based on a double-ended cold leg break is analyzed with the discharge coefficient (Ca) of 0.4. This analysis covers the whole transient period from the start of depressurization to the complete refilling of the core by using RELAP4/MOD6-EM and RELAP4/ MOD6-HOT CHANNEL for the system thermal-hydraulics and the fuel performance during the blowdown phase respectively, and RELAP4/MOD6-FLOOD and TOODEE2 during the reflood phase. A simple analytical method has been developed to account for the lower plenum filling by approximating steam-water countercurrent flows and superheated wall effects at the downcomer during the refill period. Based on the informations. at the time of EOB (end-of-bypass), the refill duration time and the initial reflooding temperature were estimated and compared with the results from the RELAP4/MOD6, resulting in a good agreement. In addition, some parametric studies on the EOB were performed. The form loss coefficient between upper head and upper downcomer was found to be sensitive to the occurrence of the spurious EOB. Appropriate form loss coefficients should be taken into account to avoid the flow oscillations at the downcomer. The analyses with the six and three volume core nodalizations, respectively, show much similar trends in the system thermal-hydraulic performance, but the former case is recommended to obtain good results.

• Nuclear Engineering and Technology
• /
• v.50 no.6
• /
• pp.829-841
• /
• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• 한국전산유체공학회:학술대회논문집
• /
• 2005.10a
• /
• pp.105-110
• /
• 2005

Evaluation of elliptic blending turbulence model (EBM) together with the two-layer model, shear stress transport (SST) model and elliptic relaxation model (V2-F) is performed for a better prediction of thermal stratification in an upper plenum of a liquid metal reactor by applying them to the experiment conducted at JNC. The algebraic flux model is used for treating the turbulent heat flux. There exist much differences between turbulence models in predicting the temporal variation of temperature. The V2-F model and the EBM better predict the steep gradient of temperature at the interface of thermal stratification, and the V2-F model and EBM predict properly the oscillation of temperature. The two-layer model and SST model fail to predict the temporal oscillation of temperature.

• 한국전산유체공학회:학술대회논문집
• /
• 2004.10a
• /
• pp.221-225
• /
• 2004

Evaluation of turbulence models is performed for a better prediction of thermal stratification in an upper plenum of a liquid metal reactor by applying them to the experiment conducted at JNC. The turbulence models tested in the present study are the two-layer model, the $\kappa-\omega$ model, the v2-f model and the low-Reynolds number differential stress-flux model. When the algebraic flux model or differential flux model are used for treating the turbulent heat flux, there exist little differences between turbulence models in predicting the temporal variation of temperature. However, the v2-f model and the low-Reynolds number differential stress-flux model better predict the steep gradient o( temperature at the interface of thermal stratification, and only the v2-f model predicts properly the oscillation of temperature. The LES Is needed for a better prediction of the amplitude and frequency of the temperature fluctuation.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.05a
• /
• pp.574-579
• /
• 1995

Saturated boiling heat transfer experiment in a hemispherical narrow space is conducted using Freon-113 to investigate an additional heat removal capability through a hypothetical gap between lower head and degraded core. The narrow space of 1mm consists of a 124mm diameter heated stainless steel hemisphere and a glass outer vessel. Within the hemispherical narrow space large coalesced bubbles are produced and these bubbles rise in random direction, causing liquid flow in from the opposite side to fill the region. Such flow in random direction makes the flow field in the narrow space very chaotic and thus enhance heat transfer. The heat transfer coefficient is higher at lower angle and at higher heat flux. The present study shows that the liquid from upper region can effectively penetrate into the gap and augment the heat removal capability through tile gap.

• Nuclear Engineering and Technology
• /
• v.53 no.11
• /
• pp.3517-3527
• /
• 2021

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Several options using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium void reactivity are neutronically analyzed in this core concept in order to improve the inherent safety of the core. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the neutron leakage under coolant expansion or voiding. The Monte Carlo calculations were used to design the cores and analyze their physics characteristics with heterogeneous models. The results of the design and analyses show that the final core design option has a small burnup reactivity swing of 618 pcm over ~54 EFPYs cycle length and a very small sodium void worth of ~35pcm at EOC (End of Cycle), which leads to the satisfaction of all the conditions for inherent safety with large margin based on the quasi-static reactivity balance analysis under ATWS (Anticipated Transient Without Scram).

• Nuclear Engineering and Technology
• /
• v.38 no.2
• /
• pp.163-174
• /
• 2006

The $Ph{\acute{e}}bus$ FP project is an international reactor safety project. Its main objective is to study the release, transport and retention of fission products in a severe accident of a light water reactor (LWR). The FPT4 test was performed with a fuel debris bed geometry, to look at late phase core degradation and the releases of low volatile fission products and actinides. Post Test Analyses results indicate that releases of noble gases (Xe, Kr) and high-volatile fission products (Cs, I) were nearly complete and comparable to those obtained during $Ph{\acute{e}}bus$ tests performed with a fuel bundle geometry (FPT1, FPT2). Volatile fission products such as Mo, Te, Rb, Sb were released significantly as in previous tests. Ba integral release was greater than that observed during FPT1. Release of Ru was comparable to that observed during FPT1 and FPT2. As in other $Ph{\acute{e}}bus$ tests, the Ru distribution suggests Ru volatilization followed by fast redeposition in the fuelled section. The similar release fraction for all lanthanides and fuel elements suggests the released fuel particles deposited onto the plenum surfaces. A blockage by molten material induced a steam by-pass which may explain some of the low releases. The revaporisation testing under different atmospheres (pure steam, $H_2/N_2$ and steam /$H_2$) and up to $1000^{\circ}C$ was performed on samples from the first upper plenum. These showed high releases of Cs for all the atmospheres tested. However, different kinetics of revaporisation were observed depending on the gas composition and temperature. Besides Cs, significant revaporisations of other elements were observed: e.g. Ag under reducing conditions, Cd and Sn in steam-containing atmospheres. Revaporisation of small amounts of fuel was also observed in pure steam atmosphere.

• Journal of computational fluids engineering
• /
• v.10 no.4 s.31
• /
• pp.12-17
• /
• 2005

A computational study of evaluation of current turbulence models is performed for a better prediction of thermal stratification in an upper plenum of a liquid metal reactor. The turbulence models tested in the present study are the two-layer model, the shear stress transport (SST) model, the v2-f model and the elliptic blending mode(EBM). The performances of the turbulence models are evaluated by applying them to the thermal stratification experiment conducted at JNC (Japan Nuclear Corporation). The algebraic flux model is used for treating the turbulent heat flux for the two-layer model and the SST model, and there exist little differences between the two turbulence models in predicting the temporal variation of temperature. The v2-f model and the elliptic blending model better predict the steep gradient of temperature at the interface of thermal stratification, and the v2-f model and elliptic blending model predict properly the oscillation of the ensemble-averaged temperature. In general the overall performance of the elliptic blending model is better than the v2-f model in the prediction of the amplitude and frequency of the temperature oscillation.

• Nuclear Engineering and Technology
• /
• v.33 no.3
• /
• pp.315-326
• /
• 2001

Direct ECC bypass phenomena that occur in a reactor vessel downcomer with a Direct Vessel Injection (DVI) system during the reflood phase of a Large Break Loss-of-Coolant Accident (LBLOCA) are experimentally investigated using a transparent l/7.5 scaled down test facility of the Upper Plenum Test Facility (UPTF). A series of separate effect tests are peformed in order to investigate the mechanisms of direct ECC bypass and to find out its scaling parameters. Various flow regimes and phasic distribution in downcomer are identified and mapped, and the fraction of direct ECC bypass is measured under a wide range of air and water injection conditions. From the counterpart test of the UPTF Test 21-D, the dimensionless gas velocity ( $j^{*}$$_{g,eff}$) is derived experimentally, which is believed to be a major scaling parameter for the fraction of direct ECC bypass. And it is found out that the direct ECC bypass is greatly affected by the spreading width of ECC water film and the geometric configuration of the downcomer.r.

• 한국전산유체공학회:학술대회논문집
• /
• 2006.10a
• /
• pp.77-82
• /
• 2006

Evaluation of the elliptic blending turbulence model (EBM) together with the two-layer model, shear stress transport (SST) model and elliptic relaxation model (V2-F) is performed for a better prediction of natural convection and thermal stratification. For a natural convection problem the models are applied to the prediction of a natural convection in a rectangular cavity and the computed results are compared with the experimental data. It is shown that the elliptic blending model predicts as good as or better than the existing second moment differential stress and flux model for the mean velocity and turbulent quantities. For thermal stratification problem the models are applied to the thermal stratification in the upper plenum of liquid metal reactor. In this analysis there exist much differences between the turbulence models in predicting the temporal variation of temperature. The V2-F model and EBM better predict the steep gradient of temperature at the interface of thermal stratification, and the V2-F model and EBM predict properly the oscillation of temperature. The two-layer model and SST model fail to predict the temporal oscillation of temperature.